1. Field
This invention pertains generally to pressurizers for pressurized water reactors and more particularly to surge line separators for integral pressurized water reactors.
2. Related Art
In a nuclear reactor for power generation, such as a pressurized water reactor, heat is generated by fission of a nuclear fuel such as enriched uranium, and transferred into a coolant flowing through a reactor core. The core contains elongated nuclear fuel rods mounted in proximity with one another in a fuel assembly structure, through and over which the coolant flows. The fuel rods are spaced from one another in co-extensive parallel arrays. Some of the neutrons and other atomic particles released during nuclear decay of the fuel atoms in a given fuel rod pass through the spaces between fuel rods and impinge on fissile material in adjacent fuel rods, contributing to the nuclear reaction and to the heat generated by the core.
Moveable control rods are dispersed throughout the nuclear core to enable control of the overall rate of the fission reaction, by absorbing a portion of the neutrons passing between fuel rods, which otherwise would contribute to the fission reaction. The control rods generally comprise elongated rods of neutron absorbing material and fit into longitudinal openings or guide thimbles in the fuel assemblies running parallel to and between the fuel rods. Inserting a control rod further into the core causes more neutrons to be absorbed without contributing to fission in an adjacent fuel rod; and retracting the control rods reduces the extent of neutron absorption and increases the rate of the nuclear reaction and the power output of the core.
FIG. 1 shows a simplified conventional nuclear reactor primary system, including a generally cylindrical pressure vessel 10 having a closure head 12 enclosing a nuclear core 14 that supports the fuel rods containing the fissile material. A liquid coolant, such as water or borated water, is pumped into the vessel 10 by pump 16 through core 14 where heat energy is absorbed and is discharged to a heat exchanger 18 typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown) such as the steam driven turbine generator. The reactor coolant is then returned to the pump 16 completing the primary loop. Typically, a plurality of the above described loops are connected to a single reactor vessel 10 by reactor coolant piping 20. A conventional pressurized water reactor typically has a separate pressurizer 22 connected to one of the reactor coolant loops that maintains the pressure of the system.
Commercial power plants employing this design are typically on the order of 1,100 megawatts or more. More recently, Westinghouse Electric Company LLC has proposed a small modular reactor in the 200 megawatt class. The small modular reactor is an integral pressurized water reactor with all primary loop components located inside the reactor vessel. Due to the arrangement of these components within the reactor vessel in this integral, modular reactor design, the traditional method for separating the pressurizer from a reactor coolant loop is not possible. Typical pressurized water reactors employ a surge line 32 connected between the pressurizer 22 and the piping in one of the reactor's coolant loops for this purpose. The pressurizer 22 controls whether coolant is added to the coolant loops or withdrawn from the coolant loops through the surge line to maintain the system's pressure within operating design limits. Thus, in a traditional pressurized water reactor nuclear steam supply system the pressurizer and the reactor are separate pressure vessels connected with a long section of pipe, i.e., the surge line 32. The surge line allows communication between the pressurizer and the remainder of the reactor coolant system which allows for fluctuation in reactor coolant system volume while providing physical separation between the warmer fluid of the pressurizer from the remainder of the reactor coolant. In integral reactor designs, a new method is required to establish the thermal separation that must be developed for the pressurizer to function as intended and also provide the surge function while being contained within the pressure boundary of the integral reactor vessel.
Accordingly, a new pressurizer interface is desired that will maintain physical separation between the pressurizer and the remainder of the reactor primary system within the reactor pressure vessel.
Further, such a new separation apparatus is desired that will provide thermal separation between the warmer fluid of the pressurizer and the remainder of the reactor coolant system.
Additionally, a new separation apparatus is desired that will satisfy the surge function while being contained within the pressure boundary of the integral reactor pressure vessel.